IR 05000213/1990013

Safety Insp Rept 50-213/90-13 on 900801-0911.Good Radiological Controls Performance Noted.Major Areas Inspected:Plant Startup Following 1989/1990 Refueling Outage,Reactor Trips on 900813 & 0903 & Safeguards Problems

ML20058A643
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	10/19/1990
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20058A637	List: IR 05000213/1990013 ML20058A633
References
50-213-90-13, NUDOCS 9010290047
Download: ML20058A643 (22)

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.-U.S. NUCLEAR REGULATORY COMMISSION t

REGION 1 o

Report No.-

50-213/90-13:

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' License.No,

DPR-61

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Licensee:

Connecticut, Yankee Atomic Power Company P. O. Box 270 Hartford, CT06141-0270 m

Eg Facility:

H_addam Neck Plant

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Location:

Haddam Neck. Connecticut Inspection:

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Dates;.

= August 1,1990 :to. September 11.-1990

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LReportings

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. Inspector:

. John T. Shedlosky, Senior Resident Inspector Inspectors:

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LAndra A. Asars, Resident-Inspector.

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iJohn T. Shediosky, Senior Resident Inspector y

~Approvedfby:.

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Donald: R. Haverkamp,; Chief Date.

-Reactor Projects Section 4A Division of. Reactor Projects--

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Inspection.Summars kInspection on August li 1990 -cSeptember 11, 1990'

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(Inspection Report No.-50-213/90-13)-

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, Areas? l'nsbected: -.Routineisafety; inspection by the-resident inspectors; L Areas,

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reviewed (included. plant startup following'the 1989/1990 Refueling Outage, c

.rbactor trips on1 August 113-and. September 3, auxi.liary feedwater: system deft-

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-ciencies,..: emergency diesel generator start failure alarm investigation',-480V.

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safeguards'busEundervoltage setpoint? problems Plant Operations Review Commit--

' teelme'etings,:and specialc and-' routine: written reports.

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EXECUTIVE SUMMARY k

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-Haddam Neck Plant

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NRC-Region I Inspection No. 50-213/90-13 i

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. Plant O'perations

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The plant' commenced power. operations for the start of Operating Cycle No. 16 on August 12.

There were two manually initiated reactor trips; both were in-4 b

response to equipment. failures.. The first on August 13, was made to allow r

W repairs to components'of the' rod control system, the second on September 3, was

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in response to a failed open steam generator feedwater regulating' valve. There j

were also two emergency safeguards features actuations; both involved the l

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automatic actuation'of the auxiliary feedwater system. They were both initiat.

-l ed by the: low-suction pressure trips of the main feedwater pumps while at 80%

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< reactor power; both of these trips occurred when-the No. I steam generator-

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feedwater. regulating valve failed full open.

In the first instance on Septem-

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'ber 2, the cause was a -personnel er.ror during trouble. investigation of a' con-I

.~ trol circuit problem; the:second, a' day later, when the valve stem failed. A

temporary waver.-of.' compliance was issued by the NRC~ Of fice of Nuclear Reactor

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Regulation on-August 30.pending issuance of a technical specification amendment

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which recognizes operator intervention following an automatic initiation of~the

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auxiliary feedwater-system.

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Radiological Controls

Good radiological controls performance was noted during this inspection period.

Maintenance and~Surveillan'ce-

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Maintenance = actions were taken to -repair a failed electricals contactor within -

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the. rod; control.fsystem and.to install a. manual isolation valveLto:back up a'

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fleaking reactor? coolant system drain header isolation *valvet Both of these

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defective' components' had maintenance activities duringathe. past refueling out-p

age. The-"A emergency. diesel; generator experienced start failure annunciations -

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'during two; tests;of August"7 and September 11'.

Although the engine started on-

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- both-occasions,.the annunciation was indicative 'ofia potential problem.

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. Security-and Safeguards?

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.GoodusecurityLperformance:wasinoted duringithis inspection-period.

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and Technical, Support ~

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f 1 Failure a~nalysis of theLNo.1 steam generator feedwater regulating valve was 1,'

'made along with-a. technical-evaluation of?the short term effects of-power.

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operation with loose valve parts'within'the'feedwater system.

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? iscovered" potential problems with the 480 volt safeguards bus undervoltage

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relay"setpoint'which may have affected_the ability to initiate post accident

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sump ~ recirculation.. It also discovered that the need for operator interven--

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ttion fol. lowing'an. auxiliary feedwater automatic: initiation had not been -recog-

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nized in the1 plant licensing ba' sis.- Additionally.. engineering reviews discov-

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ered that although the-reactor coolant' system loop isolation valves were

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1 credited for mitigating ~a steam generator tube. rupture accident,-the torque:

switch; settings were-incorrect, l

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' Safe'ty Assessment and Quality Verification I

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' Although licensee-action in response to the above referenced problems was-

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Lgenerally prompt.and thorough, the: emergency diesel generator start failure

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problem of August 7 was-not pursued aggressively and reoccurred during subse -

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quent surveillance-tests on September 11.

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TABLE OF CONTENTS e

Page

1,z (Summary 'of Facility Activities

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Plant Operations (71707, 71710, and 93702)*.......... 3 F

2.1~-Operational' Safety Verification

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L 2.2-Engineered Safety Features System Walkdown........ 3 2.3' Follow-up.of Events Occurring During the Inspection

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' Period.........................

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~ 2.3.1 Reactor Trip on-August 13............. 3 i

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. 2.3.2 Reactor Trip on September _3.............

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- 2.3.3. Temporary; Waiver.of Compliance from Requirements

. for Automatic Initiation of the Auxiliary

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Feedwater System............

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j 2.3.4 ! Actions to Correct Reactor System Drain

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Val _ve Leakage

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Rad'iological/ Controls.(717'07),................

'7-l 4.

Maintenance'and ~ Surveillance (61726, 62703, and 71707')

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4.~1 ' Maintenance Observation

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4.1.' 1.

Failure;of the No. 1. Steam Generator Feedwater

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Regulating ) Val.ve a...

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. Surveillance Observation.....

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- 4.2.17 Emergency-Diesel Generator' Start Failure

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Ensineering: andLTechnical; Support -(37700, 37828, and 71707)1.c.

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6.-11 Discovery of LPotential' Probism with:.480 volt; Safeguards

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ge % 90712,Landc92700)_..1.,........;....--...

< <- c- .- . -.. -,.. , b <7~.-11_ Plant Operations ~ Review Committee and Nuclear ' , yi . J7.2L Review ofLWritten Reports......_...........

p~* _ _ _ -. fReview Board .o.. . -..

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Exit Interviews (30703)

. 17' . -...-..........-..... g --c eW - M. A --*:The NRC' Inspection Manual inspection _ procedure-or temporary instruction:that Lwas used as inspection guidance.isL11sted for'each applicable report 'section... ,=.2 +

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p ~ DETAILS s 1.

Summary of Facility Activities At the_beginning of the period, the reactor plant was in Operational Mode-3, Hot S_tandby.

Functional full flow testing of the auxiliary feedwater system had been completed.

, -The reactor was cooled to 200 degrees F, Operational Mode 5, on August 3, to allow the installation of a manual isolation valve within the reactor coolant system low = point drain system.

This action was needed to correct through-seat leakage of a motor operated valve.

On August 7,.the "A" emergency diesel generator failed to successfully meet the test acceptance criteria for a semiannual fast start surveillance test required by technical specification 4.8.1.1.2.4.

The engine was observed to accelerate.to 900 rpm a 12 seconds versus the required 10 j seconds or'less.

Causes for the start-. failure could not be determined; the Diesel Generator was successfully surveillance tested and returned to-service the same day.

i A reactor plant heatup was conducted following drain line modifications and the reactor was made critical on _ August 12 at 8

53 a.m. to begin b

Operating Cycle No. 16.

, A manual reactor trip was performed on August 13 at 12:20 a.m. when it was decided that a reactor shutdown was required to repair-arcing electrical L ' components within the reactor rod control system.

Because the arcing L< occurred during normal. control rod motion, the licensee determined that

1 ' l driving control rods.in 'a normal manner _was unacceptable. A manual - ireactor trip prevented any additional: damage which ~might have occurred through the_use ofJthe, rod control system.- Following repairs the reactor o ' y 'was again'made~ critical at 4:33 a.m. on! August 13.

' Low power physics testing was. completed on August 14; On-AugustL15.at- , ,7:'30 a.m.' the main' generator _was phased to the grid,'and at'11:17 a.m'. the the plant entered Operational Mode 11.

Between A' gust 17'and. August 30, the reactor remained below ten percent u p rated power pending1 resolution'of two outstanding issues concerning the s . auxiliary feedwater ( AFW)> system.- On August 30', NRC issued a. temporary-l

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' ' - ' waiver of compliance to-Technical Specification (TS); Table. 3.3-2',-Item 3 l ,

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P " Auxiliary Feedwater" allowing the plant to proceed;in power ascension o '(above.10; percent' power)iwith AFW' automatic. initiation-system not meeting d > h (the1TS operability" requirements. :The_ associated proposed TS amendment-j

redefined the'o'perability of automatic AFW initiation'.to include 1 credit: i - forcoperatorcactionito adjust"AFW full flow' and-also allowed reliance'on

' ~ ' theinon-safety; grade control' air system to ensure successful automatic AFW . initiation;

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r.. , , ' .On September 2 at 9:26 a.m., an emergency safeguards feature (ESF) actua-tion -occurred at 80*; reactor power in that AFW was automatically initiated when both main'feedwater pumps tripped on low suction pressure. A feedwater flow excursion to the No. I steam generator resulted when the

No. 1 feedwater regulating valve opened fully during control circuit i a troubleshooting, Personnel had failed to place the No. I feedwater regulating valve controls in manual prior to opening a portion of the - electrical circuit.

However, a combination of both quick operator re- ' sponse and automatic restarting of the main feedwater pumps prevented a serious steam generator water level transient.

The reactor remained at-80*; power.

, However, the reactor was manually tripped from 80*; pcwer on September 3 at

A 4:57 a.m. in response to high level in the No. I steam generator along with the inability to control the corresponding feedwater regulating valve, Level had reached 85*; of the narrow range indication.

The AFW-isystem automatically initiated and steam generator level was maintained , using the:feedwater regulating valve bypass valves.

The motor-operated.

main feedwater regulating valve isolation valves were shut, The No. 1- . feedwater regulating; valve was disassembled and found to have _ failed, its

stem broken and'several' pieces of valve material missing.

These were ' .apoarently carried.away by the feedwater flow which was in excess of 3E+06 . -pounds mass-per-hour, , 'The' reactor was made critical at 7:11 p.m on September,3 while repairs [ '

to the-feedwater regulating valve were in progress. The investigation

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- .into the failure was c'o.nducted during repair process.

This included a - modification'in which a' seal-weld was.m'ade to prevent unthreading the , valve disc from_the stem.

Based on th'e results of the licensee's failure ' anal'sisLand a technical evaluation of the-potential damage caused by the - ' E y + -unretrieved11oose parts, the main feedwater. system was placed-in service' ' _ cand the-turbine generatorgsynchronized with the' grid at 10:55 p.m. on.

,e' ' September 8.

The loose parts technical evaluation;was approved by:the si . . licensee to allow' power operation'forXonly seven days. A-final: evaluation , Jwas due.by, September 15.

L Discussion of this' issue is carried over -to the.next inspection period.

The reactor was at 80*; power at the' conclusion of this period ~ s . 'The' inspection activities during-this report per:od included 139 hours-'of

. inspection'during' normal utility working. hours..In addition-the review of-s , ' plant operations w'asL routinely conducted during: portions of backshifts - ,% (evening shifts)- and~ deep.backshif ts (weekend:and night shif ts), Inspec-tion coverage was provided for.35 hours-during backshifts and 8 hours U, . ~ ,during~ deep backshifts; il J.

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, - 2,q P1 ant Operations-2,1 Operational Safety Verification ' "' The. inspectors observed plant operation and verified that the plant ! was_ operated safely and in accordance with licensee procedures and t y regulatory requirements.

Regular tours were conducted of the i following plant areas: ! I control room security access point --- -- -- ' primary auxiliary t,311 ding protected area fence -- -- intake structure ! radiological control point -- -- - n, ' electrical switchgear rems diesel generator rooms

-- -- auxiliary feedwater pump rovin turbine building -- -- i L There were no'noteable observations identified.

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2.2 ~ Engineered' Safety Features System Walkdown [ In; addition to routine observations made during regular plant tours, - , u, -the inspectors conducted walkdowns of the accessible portions of ' ' ' selected safety.related systems.

The inspectors verified-system operability through reviews of valve lineups, control room system prints, equipment conditions, instrument' calibrations, surveillance

test frequencies-and.rssults, and control room indications, During ' this; inspection period, walkdowns of the following systems were performed: ' e " Emergency Diesel Generators ' - - - -T ~ Auxiliary Feedwater'

-- There were no noteable~ observations identified.

2.3: - Follow-up of Events ' Occurring During Inspection Period- . . . -During the inspection period,:the inspectors provided on-sitei l ~g ' coverage land; follow-up of' unplanned events. ~ Plant conditions,- - -v . -alignment of safety systems c and; licensee. actions were reviewed.' The Tinsoectors confirmed thatErequired.' notifications were made toithe I NRC. During event' follow-up, the-inspectors reviewed the-correspon-1 i.

ding plant;information ' report (PIR)1 package, L including the:evente s < e,1 details,: root cause analysis, and? corrective' actions taken to preventi 'Li , -recurrence.

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, ._ '2.3.1-Reactor Trip on/Augilst 13 <

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m [Y A manualireactor trip.was performed on August 13 at 12:20 a.m..

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whe'n it was-decided that a' reactor shutdown was requi' red to ' . repair arcing _ electrical components within the reactor rod

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gn E !: k, control system. The reactor plant was at normal operating temperature and pressure and an approach to reactor criticality' in progress, .

, Because the arcing had-occurred in components associated with normal' control rod motion, the licensee determined that a , manual reactor trip would prevent any additional damage which might have occurred through the use of the rod control system.

! Following repairs, which included the replacement of a damaged i electrical contactor, the reactor was again made critical at-4:33 a.m. on August 13.

a-The-trip had occurred while the reactor was subcritical during the second approach to criticality for low power physics' test- ' ing. Operators responded to a process computer alarm which , indicated that there was a difference between bank average and individual rod positions.

They found that with bank "A" rods , at 215 steps, control rods No-4 and 5 were.at 188 and 189 . = steps, respectively.

Recovery procedures were not effective

in moving the two control rods.

! . An inspection of the rod cont'rol circuit cabinets, located in v the'"A" electrical switchgear room, found that a power relay , used to apply:lif t coil _ power was arcing excessively and had ' probably failed.. Although this. component failure resulted in the inability to withdraw the two: control rods, other portions

of the circuit-provided power to= hold the control rods in position.

lThe licensee determined that-the rod control cabinet needed to ' be'deenergized in order.to properly replace the_ failed compo-a W nent<and to inspect 1the entire cabinet.for debris. They were-w concerned'that the operation.of' components within the cabinet ~ ? ' during cont'rol rod insertion might;have caused additional .

t damage-Because of these concerns? - a ' manual reactor : scram was ~

. ' ' fn .used for shutdown; ,' ' 'A? defective electric /contactor associated with the control ' ' ' r rod lif t coil power circuit. for the twol rods was the source of

~theiproblem.

The'contactor. arc chute had come loose-allowing ' m momentary contact between its' moving contact and the arc chute

g attaching device.- This resulted in an arc spraying debris

< i' within5the rod-control cabinet.

The contactor was a Westing-

house; Type MM'420Lrated at 126 volts'd.c; The' defective (~

contactor.was replaced, ana the rod'controlLcabinet was , inspe'cted'.to verifycthat simil~ar arc chutes were retained J !

properly and that-there was no consequential damage from

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. This and other rod control components had been cleaned and ' M inspected during the refueling outage (Reference preventive , a maintenance procedure.PMP 9.5-2, Rod Control System Preventive Maintenance, Revision 7, dated April 5, 1990).

The event may have been due to the failure of the arc shoot retention device o or to the failure to verify proper engagement of the retention b device.

, The reactor was made critical on August 13 at 4:33 a.m.

Low power physics testing was completed on August 14.

The . inspectors.had no further questions.

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1) m 2.3.2 Reactor t' rip on September 3 _ The, reactor wast manually tripped from 80*4 power 'on September 3 > J at 4:57 a.m.' in-response to high water level in the'No.1 , ' . steam generator along with the inability to control _the corre- "sponding feedwater regulating valve.

Level had reached 85*4 of- > theinarrow range indication; The AFW system automatically . initiated and. steam generator level was maintained using the feedwater regulating valve bypass valves.

The motor operated .,,

. main feedwater regulating valve isolation valves and the No. 1- _1 , , -main steamctrip_ valve were shut.

.The No.1 feedwater regulating valve was disassembled and found:to have-failed..Its stem was broken and several: pieces 'q , of valve material were missing, apparently carried away by the ' s-feedwater flow which was11niexcess of 3E+06 pounds. mass.per.

m .- -hour.. q

' 4 sin a'ddition_to.the replace ~ ment.of iheifailed' valve stem'and.

' sdisk; assembly, the licensee. implemented a design change in hl , 'which theistem wasiseal welded to the valve discHto prevent ' m.

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unthreading w Valve; repairs 1were. completed total

low the: main

' feedwater. system _to be placed in service and:the turbine ' T g~ generator' synchronized with the grid-on. Sepember 8.

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analysis:ofitheyalve_ failure was completed'along with a

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• technical-evaluation to allow power operation fory seven days..

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, x . A safetyjanalysis which justified. power operation through. the j[f ,end of theToperating cycle was5 completed:during the next'resi-1 . g ' dent i_nspection peri _od.

The licensee was actlyely' pursuing.

s . < . resolution :to the: above mentioned issues.

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, 2,3.3 Temporary Waiver of Compliance from Requirements for Automatic ,_ Initiation of the Auxiliary Feedwater System ' During this inspection period, auxiliary feedwater ( AFW) system surveillance testing revealed system vulnerabilities to losses of control air.

An engineering evaluation identified the necessity for operator action to attain sufficient AFW flow following an automatic initiation.

This does not support the requirements of Technical Specification (TS) Table 3.3-2, Item 3 " Auxiliary Feedwater" which requires that automatic initiation of AFW be operable when the plant is at or above ten percent rated power.

The licensee submitted a proposed TS change and request for a temporary waiver of compliance which redefines the operability of AFW, The redefinition includes credit for operator action to attain full AFW flow and reliance on the control air system for automatic AFW initiation. The temporary waiver was granted by NRC's Office of Nuclear Reactor Regulation on August 30 pending issuance of an emergency TS change.

Two conditions accompanied this approval; an operator must be dispatched to the AFW room when an automatic initiation of AFW occurs to verify that the pumps are operating as expected; and the licensee must provide an AFW initiation event analysis with best estimate assumptions in accordance with ANS 58.8, " Time Response Design Criteria for Nuclear Safety Related Operator Actions", and perform a walk-through of the process to verify thatLoperator actions can be performed within the required response times.

This= issue is discussed in more detail in section 6.2 of this report.

2.3.4 Actions to Correct Reactor System Drain Valve Leakage A motor-operated isolation valve (DH-MOV 562) was found to = have significant through-seat leakage when the reactor coolant system was pressurized to normal operating pressure

(2000psig). 'The valve. leakage allowed water to enter the ' reactor plant drain header from'the pressurizer surge line.

This condition pressurized the drain header and the drain cooler heat exchanger to primary plant pressure.

On August 2, the plant was cooled down for repair of DH-M0V-562. The licensee had determined that the valve was binding internally preventing its full closure.

Corrective action was desired to eliminate the potential radiological problems caused by the existence of slight s leakage of reactor coolant into the component cooling water ' ____w___ - _ -a-_----_x -_ x --

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r 4 , g system through the drain cooler.

The reactor was placed in C-Operational Mode 5 on August 3 at 6:30 a.m.

The licensee elected to accomplish the repairs by installing a

new manual isolation valve down stream of the leaking motor-operated valve. This process included the use of a freeze seal on the one-and-one-half-inch drain line.

This was d decided upon because the disassembly of DH-MOV-562 would require-draining of the reactor vessel to the center line of the vessel hot leg.

The inspectors reviewed the freeze seal procedures including , the safety evaluations made to establish the seal location, ! emergency actions,to address seal failure and non-destructive

inspection of piping following the placement of the freeze seal.

The design change package was found complete and included the seismic analysis of the-drain header modifica-tions.

There were no unacceptable conditions identified.

A one-half-inch sample line is connected to the drain line -- r

adjacent'to OH-M0V-562.

Because the sample line configuration effectively bypassed the new manual isolation valve, the -; sample line was_ required to be cut and capped. This was also accomplished by'using-a freeze seal for isolation.

' The.cause for the failure of valve DH-MOV-562 to fully close t y cannet be determined.until it is disassembled during;a future'

!

outage.. However, its stroke was measured.and found to be less ~ ' ,than.its normal 1.5 inches._ ~ Radiography of.the valve body ' failed to_ identify a: foreign object within the valve which P would interfere.with its closure.

There is a possibility that [ the disc was' reversed when' installed intthe valve ~ guides.

' , gg > - ' c .The inspectors had'no further questions.

. . o 'a 3;' Radiological Controls , o p

During routine-inspections of the. accessible plant' areas,'the inspectors

~ % observed the;implementationcof selected portions of the licensee's radio - ' logical controls. program.

Utilization and. compliance-with ' radiation' work" _, permits (RWPs) were reviewed:to ensure that detailedEdescriptionsL of

, ' , g Lradiological conditions were provided and that personnel adhered to~RWP f requirements. -The -inspectors observed controls of access to various. 4 radi_o~1ogically controlled areas and-the use of personnel monitors and , ' i' cfrisking methodslupon exit from:thosefareas. LPosting!and control of-M'

radiation. areas, contaminated' areas and hot spots,

and labelling;and

l control of containers holding ' radioactive materials"were verified to be in i .accordance with' licensee. procedures.

Health physics ~ technician control

, S and monitoringfof plant activities were determined to be adequate.

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? 4 '. Maintenance and Surveillance.

, 4.1' Maintenance Observation i The inspectors observed various corrective and preventive maintenance activities for compliance with procedures, plant technical specifica-tions, and applicable codes and standards.

The inspectors also verified the appropriate quality services division (QSD) involvement, r use of safety tags, equipment alignment and use of jumpers, radio- ' logical and fire prevention controls, personnel qua'lifications, post-maintenance testing, and reportability.

Portions of activities that were reviews' included: l n-i Troubleshooting of "B" emergency diesel generator electric fuel } D -- oil pump-failure to start; Installation-of manual isolation valve-to back up leaking drain -- header isolation valve,. DH-MOV-562; . - d Installation of a.line cap on the pressurizer drain sample line;. -- -Repairs to leaking' isolation valves associated with pressurizer -- level instruments; Investigation into apparent air binding-of-the charging pumps; '-- q

' Replacement of failed Type.MM 420 electrical contactor within: rod -- . control. system; ,, l Investigation into failure of: No.1-steam generator -feedwaters --: Eregulating valve; Inspections of. Nos. 2,' 3.and 4. steam generator 'feedwater regulat-

- -

ing: valves; and, E 4-. Investigation into the lower detector failure of Nuclear Instru - 'i mentation Channel No.' 1.

,s There were.no noteable observations identified, j , ..

f L4.1.11 . Failure of No. I Steam Generator Feedwater Regulating Valve e v ' :w " ' a ' Following: the September 3 failure. of-the No.1 steam generator.

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. fe'edwater r'egulating: valve, disassembly revealed that-the.

h] g ' valve stem;had-broken above'the point where it was threaded-O ,intot theLyalve _ disc. : The~ disc had fallen to:the; bottom of-the e valve body, fully. opening-the valve flow-stream. Additional. ~ 'ly,.the threaded portion of the stem was missing along with

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several small pieces-of valve seat material.

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, e ' ' ,% All;four feedwater regulating valves had been removed from the i, system, shipped to the vendor, and refurbished during this past refueling outage.

They_were returned as complete assem-blies, and were installed without disassembly.

" Following the valve failure the licensee discovered that the threaded stem and disc assemblies were not seal welded as was ., [jb the convention in the past.

The vendor had apparently modi- ' fied the design to stake or pin the stem to the disc.

6 ' The' licensee's failure analysis concluded that the stem failed

to remain engaged with the valve disc due to severely damaged.

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threads and a-sheared locking pin. As the disc became j unthreaded system flow increased vibrations which resulted in m fatigue fracture of the stem.

The spool shaped disc was spun in the valve body allowing the stem fragment to come loose and be carried away by the very high flow.

i The analysis endorsed a course of action to inspect-and modify I all four valve stem assemblies. The stem and disc were inspected-by radiography to insure that the locking pin was fully inserted through both the disc and stem pieces. The assembly was then modified by making a seal weld between the ,

disc and stem to prevent unthreading.

This was the same ,

c welded configuration of the previous design.

] The inspector reviewed.the licensee's' investigation and j repair.

In: addition to-the work in progress, this review-i e included the component fail _ure analysis (Reference No.. PSE-CE- ^ 90-520), the valve modifications (PDCR 1005) and the-technical evaluation (Reference'No, PSE-EM-90-321) concerning the return to power operation with unretrieved~1oose valve parts.

j Licensee actions were considered acceptable, y a _The loose parts technical evaluation was approved to allow power _{ ,, , ' ~ op_eration for only 'one week. Atithe end.of this inspection - j 7!,' period, a safety analysis was under preparation to justify ' operation through'the end of this operating cycle with these . parts 1within the.feedwater system. This will be addressed .] during the next"residentVinspection report.

l g-.,p s y m - $ pu 14t.2 Surveillance Observation-l y a f The inspectors witnessed; selected surveillance tests to determine j , whether: Lproperly approved procedures were in use;. plant technical e . specification frequency 'and action statement' requirements were f satisfied; necessary equipment tagging was performed; test instrumen-i rc " ', .tation was in calibration and properly:used; testing was' performed by J . l qualified personnel; and test results-satisfied acceptance criteria-- j +' ' 4'~ or were properly'dispositioned.

Portions of activities associated f +

with'the following procedures were reviewed: l w [f3',1

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' ! < ' i -- SUR 5.1-141, " Functional Test for Auto Initiation Scheme for ' ' Auxiliary Feedwater"; ' -- SUR 5.3-6, " Control Rod Reactivity Worth Measurements"; l r . -- SUR 5.1-17A(B), " Emergency Diesel Generator EG-2A(B) Manual , , Starting and Loading Test".

, , - There were no noteable observations identified, e 4-2.1 Emergency Diesel Generator Start Failure . On August 7, with the plant in cold shutdown, the "A" emergen- '

cy diesel generator-(EDG) experienced a start failure during

% routine surveillance testing.

The six-month surveillance, SVR j $ 5.1-157A, " Emergency Diesel Generator EG-2A Fast Start and

b Load Test",.was being performed using the west bank of the air , start system'and the excitation panel to start the engine.

, ' ' The EDG started, however, ito failed to reach the full speed of- -! -900 rpm within 10 seconds as required by Technical Specifica- ' . tions_(TS)4.8.1.'1.2.

The EDG started in 15' seconds and f ' ' received a' start failure alarm.

A start failure alarm occurs.

q ?when either the EDG fails to reach 40 rpm within three seconds' ~ or:100 rpm within four seconds.

The EDG was subsequently c.1 shut down and another start was attempted. On this second' - ' start-the.-engine attained full speed within the required 10 seconds, however, start failure and circuit malfunction alarms.

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were. received, indicating a: sluggish start.

' ' ' ' .' The "A"-EDG was shut-down and declared inoperable;1 operability' W of,the:"B" EDG was verifled.. Troubleshooting.of the "A" EDG

start. failure:and circuit-malfunction,were initiated.. Causes for theselalarms could'not be determined.

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' SUR-5.1-157A Nas performed' a thi rd-time on Aug' st 7 as-a' ' u ' retest following troublesho'oting.

The-EDG started:satisfacto- ' - . ily, n'oltrouble, alarms were received and.the engine was ' ' r n] subsequently. declared operable.- -This,was-thejfirst start failure in the last:100 valid engine? t , tests'..The licensee submitted a special report, dated Septem-J ber'4.-1990e in accordance with TS-4.8.1.1.3.

The inspector j N reviewed the reportland; verified:the-information to be

, accurate.' <

.On September.1.1, with the' plant at about 80*4 power,:the "A" EDG experienced another: start' failure alarm duringx routine.

- testing.

During the:mo'nthly surveillance,'SUR 5.1-17A, j , " Emergency Diese1' Generator-EG-2A Manual Starting and Loading: ) Test", the licensee attempted to: simulate the August 7' failure ,

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, h - p by: selecting the air start system west bank and starting the 'V engine from the excitation panel.

The engine started as required but a start failure alarm was received.

The engine was shut down and subsequently restarted, loaded, and success-p fully run for the monthly test.

. At the conclusion of the inspection period, the licensee had.

! not determined the cause of the start failure and circuit i t.ialfunction alarms.

The operability of the "A" EDG has been .' ' verified by successful performance of the TS required surveil-lance tests.

Preparations were being made for further testing , ... - of the "A" EDG with monitoring equipment connected to the W.

start circuitry. The inspectors'will follow this testing j M activity as part of the routine surveillance observations.

' 5.

Security- ~ During routine inspection tours, the inspectors observed implementation of

portions of the security plan.

Areas observed included access point.

I search equipment operation, condition of physical barriers, site access " control,- security force staf fing, and response to system alarms and ~ degraded conditions.. These areas of program implementation were deter - m m_ined to be adequate.

- 'l 6.

' Engineering and Technical Support

a The.. inspectors reviewed selected engineering activities.

Particular

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attention was given to safety evaluations, plant operations, review commit-j

' 7 L tee' approval' of modifications, procedural controls, post-modification j '* . testing,: procedures,. operator training, Land UFSAR and drawing' revisions.

<! 6.1 ' Discovery of Potential problem with 480 volt Safeguards Bus ~Dndervoltage Setpoint [ 'As'a result of=an engineering review on August 2, the licensee ' . -found a. potential problem in the implementation of-the-emergency E* operating procedures for~the post accident 1 containment recirculation l phase, lhe= reactor was in'0perational Mode 3,LHot Standby, at the , , ,WF LtimeEof the finding.

, . . J . 1The: transfer to sump' recirculation is. accomplished.after the_ vessel.

sS.fu injection phase by usingia ' residual. heat ; removal (RHR) pump to supply-

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> A; suction pressurento a'high pressure safety? injection-(HPSI) pump..~ Ati.

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, L.that time the low pressureisafety injection (LPSI) pump would'have j g

been shut down and aLsecond service water' pump andsRHR pump started a

.by opeiate actions. This was implemented through emergency operat-e ing procedure E0P;3.1-0, ES-1,3, Transfer to Sump Recirculation.

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3; " I 12- ' . , Theiconcern was that the emergency diesel generator (EDG) voltage ' droop was large enough to actuate a bus undervoltage relay when

starting a HPSI pump.This relay would then initiate an automatic i trip of the RHR pump electrical breaker on safeguards 480 volt bus 5 [ or bus 11.

' The relay is part of the load shedding scheme and operates when the , 4160 volt safeguards bus, and therefore its 480 volt bus is deenergized. The time dependent undervoltage detection instru-mentation, referenced in technical specification 3.3.2, monitors the

4160 volt bus.

Relays 27-5 and 27-11 monitor the voltage on 480 volt bus 5.and 11, respectively. Their operation will cause a trip of . C certain electrical load breakers, ' The potential for trip of the RHR pump breaker was of concern because-u in the worst case scenario, the loss of suction pressure to the'HPSI-pump may have caused pump damage or pump failure.

U-Although-the HPSI pump motor is powered from the 4160 volt _ safeguards bus, the 480 voit bus voltage profile caused by starting-the HPSI . pump-during this scenario was calculated'to be 70'e of normal for , eight seconds. The 480Jyoit bus undervoltage sensing relays had-been a . set.at 70*4'of normal voltage.

--The licensee's corrective actions were based on the 4160 volt time N' dependent undervoltage system providing safeguards system protection, i

The'480 volti undervoltage. relays carried out the load shedding-for s

y* certain equipment'. The undervoltage settings of these relay were oq changed on= August-3 from 70's of-normal voltage to' 50*; of' normal, p

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InLaddition.to changing the relay setpoint,. engineering reviews-lwere made of::the emergency operating procedures (EOP) to confirm that- 'the^ electrical bus voltage profile problems offthis' issue were not ,, ! V: present in'other procedures.

Also, the; licensee has-committed to deeper involvement 1 by' the. appropriate engineering disciplines in;the g ' multi-discipline reviews supporting theLintegrated safety evaluations W' made.for future E0P' revisions.

There is consideration to maintaining h, .a summary of the emergency. diesel' generators load profiles.ont site, d " LA safety' evaluation, No.tISE/CY-90-061, was completed on August'6 to > 3' assure thatLin;the! transfer to sump; recirculation or two-path' recirculation the. emergency diesel generators were not loaded beyond - ., t

their; ratings and that the: required core. cooling equipment'wasi . hp available.- ' . f 'y' . ,

' LThe inspector followed the licensee's actions including the:setpoint lchangefprocess'and'the above referenced safety evaluation.- The , inspectors had no further questions, , , u

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' 6.2 Operability Requirements for Automatic Auxiliary Feedwater System Initiation During this inspection period, the licensee became aware of two auxiliary feedwater (AFW) system vulnerabilities which prevented full operability of the AFW automatic initiation system and consequently restricted plant operation to less than ten percent of rated power.

Specifically, the AFW system operability is dependent upon Se non-safety grade control air system and operator action.

. . The Haddam Neck Plant AFW system contains two, full capacity, steam L " driven AFW pumps.

Each pump turbine has an' air operated steam admission valve which is designed to fail open on a loss of the control air. system.

Each steam admission line also contains a relief valve which is set to lift at 650 psig. When an automatic AFW , . initiation signal is received, the steam admission valves open-partially to admit steam at 450 psig to the turbines. Adjustments of ~ z-AFW flow can be made manually by-repositioning the air-operated steam admission valve. The design basis AFW flow to the steam generators ' is 268-gpm, , In September, 1989 (the beginning of the 1989/1990 refueling outage), , the licensee conducted flow testing of theLAFW pumps.

The "B" pump np ' , . passed the surveillances; the "A" pump did not develop adequate flow.

(See NRC Inspection Report 50-213/89-12 for additional details.) The-l'A" AFW pump was sent.to the vendor for overhaul and mock-up testing iduring the outage.

During the-previous inspection period (See NRC~ Inspection Report

.

50-213/90-12),' the licensee conducted retests of.the AFW system.

The flow tests indicated that both pumps : produce adequate flow when- , operating at" full' design speed.

Howev'er, revised. system modeling.

, -calculations which _take into' account system losses and pump degrada-

tion have concluded.that, with the steam pressure following automatic- , Linitiation (450 psig),;the AFW pumps. provide only about 240 gpm to: -; the steam generators.. 'This,is not ' sufficient flow to meet the' design' ' basis _ requirements without_ operator action within four minutes.

Furthermore, during-system testing on August 12, both_AFW turbines-tripped on overspeed:when AFW was. automatically initiated.

Plant-c g-operators and engineers witnessed this testing.and noted that, when-l ~ the steam admission ' valves-opened, the= relief valves-liftedLand steam-Q ' o pressure was'. greater than 600-psig.The resultsLof this testing' _ '; . ' indicate:that for ~ events (e.g. seismic: events) which result in rapid - " steam admission of. greater than 450 psig,- the AFW turbines lcan trip- 'on'overspeed and therefore be unavailable without manual ~ action to ~

' resettthe=overspeed trip mechanism, a- < ^k.

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i .> . p . Based on these deficiencies with the AFW automatic initiation system, ! u-the licensee concluded that under certain condi. ions AFW supply may ! not be sufficient to remove residual heat. This was determined to be i reportable to NRC in accordance with 10 CFR 50.72(b)(2)(iii)(B); the i notification'was made on August 20.

Several conference calls were held between the licensee and NRC to d_iscuss the;e issues. Aspects discussed included: system design i basis and operation; Probabilistic Safety Study assumptions; associ-

ated emergency procedures; operator training; and system failure i , ' modes.

' + -The licensee submitted to NRC a proposed emergency change to Techni- . ' cal Specifications (TS) for AFW automatic initiation on August 25.

This request was supplemented by additional information and a request i .for a temporary waiver of compliance, dated August 29.

r E-0n' August 30, the NRC Of fice of Nuclear Reactor Regulation granted the temporary waiver of compliance from TS Table 3.3-2 Item 3, " Auxiliary Feedwater", thereby removing the power ascension.restric-j 4. tion.for operational cycle 16.

In support of this TS change, the ' , licensee has committed to dispatch an operator to the AFW room o > . whenever an automatic AFW initiation occurs and to provide an AFW , initiation event analysis. The event analysis will use the guidance for best estimate assumptions found in ANS-58.8, " Time Response , y Design. Criteria for Nuclear Safety Related Operator Actions."

F . Additionally, the licensee will perform a walk-through of the AFW initiation-process to verify that operator' actions can be performed- ~ within the required time.

' , ',The1 inspectors participated in the conference' calls and, reviewed the-

,p materials submitted to the-NRC for accuracy.

Several Plant Opera-

l tions Review; Committee'(PORC) meetings convened for consideration of-J-this matter were attended as we'll as a joint-PORC/ Nuclear Review _ L Board meeting.

The--inspectors will' review licensee. actions to comply-

• with-the conditions of-the temporary waiver and TS change during-

!< future inspections.

[ 6.3: Incorrect 0-Ring Materiil Installed in Valve Motor Operators

LThe licensee discovered that.an incorrect 0-ring,materialiwas-used in ' four valve-motor operators. :The material' which was not suitable'for-s , fuse with: petroleum: products,- was installed ^in locations' serving as: , ' ~ grease seals., This error occurred during the19871 refueling outage j ?. and involved 1 Crane Teledyne model T-4 actuators.

i " J .The discovery was made through an August'6. engineering evaluation j , % af ter. discovery.of swollen seals which were binding' the movement of- > ' the~ torque-spring and worm g' ear assembly.

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t-The vendor had supplied an ethylene propylene (EP) elastomer as a

replacement for the original Parker N-572-6 compound which was no l longer available, This material was purchased in 1986 and used the in maintenance during the 1987 refueling outage without an evaluation ' of the material substitution.

The replacement material was used in four valve motor operators: BA-MOV-32, BA-MOV-373, BA-MOV-366, and CH-MOV-257.

These are valves associated with the transfer of charg-

ing pump suction path from the volume control tank to the refueling water storage tank, > -The effect of the swollen 0-ring seals was to hinder spring pack deflection thereby yielding higher operator torque values.

However, there was no observed degradation in the motor operators, , 'The licensee's corrective actions have included the replacement of a the incorrect EP 0-rings with Viton 0-rings and the removal of all EP

0-rings from stock,

s-The inspectors had no further questions.

" , , g '6.4. Reactor Coolant Loop ' Isolation Valve Operator Torque Switch Settings ' [ > The. licensee di'scovered on August 6 that torque switch settings for, reactor coolant. system (RCS) loop isolation valve motor operators was .less than that recommended by the manufacturer, These valves have " been' considered by the-licensee to be an alternate method to stop a ~ ' release of radioactive material in the event of a steam generator r tube' rupture (SGTR)-accident-

The manufacturer had specified torque switch' settings for. loop / isolation valve motor operatorstto operate.with a 500, psi.differen- '% tiall pressure, Upon... inspection all eight valve motor operators were

found with^ settings less than the(required 2

7/8, As found values'

~ - . E ranged between 2 3/4:and 1 3/4;.the lowest: expected to close against t 'a' differential pressure'of only'260 psi, , o 1There had been:an in' adequate control of these swit0h:setpoints'in . s part'.because credit lwasi not taken.for the valves. to> mitigate the

, effects offa!SGTR accident, However,Lthe-original-plant design basis-m

for these; valves wereJas: maintenance. isolations.

A Currentcanalysis~ credits;the use'of'the valves for!the isolation of' ' ' J' ' ~ t an RCS loop :and :the ' damaged steam: generator inlthe event the RCS- < D ~ could not be: cooled a'nd.depressurized. The Updated Final Safety ' T - Analysis : Report (UFSAR), assumes that isolation procedures may be: q ' y ' 'W accomp1_ished'within 30 minutes of a"SGTR accident tol comply with'the requirements; of?10LCFR Part'100 for of f site radiological dose ' , ' limits! 'In the event.that a release to the. atmosphere was occurring-y~ " nX E,from the damaged steam generator,-the-RCS loop isolation valves may;

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, r be shut at 400 psi primary to secondary differential pressure.

This event is addresstd in UFSAR Section 15.2.10, Steam Generator Tube Rupture.

The licensee has adjusted the valve motor operator torque switches tc the manufacture's recommended setting and included these setpoints in the setpoint control program. This program contains data for other ' safety related vaive motor operators.

L The inspector retiewed the actions taken; licensee actions were considered,to be adequate, 7.

Safety Assessment and Quality Verification , ,, 7.1 Plant Operations heview Committ<_ and Nuclear Review Board The inspectors attenood several Plant Operations Review Committee (PORC) meetings and a meeting of the Nuclear Review Board.

Technical - specification 6.5 requirements for required member attendance were . verified. The meeting agendas included procedural changes, proposed changes to the Technical Specifications,. plant design change records, and minutes'from' previous' meetings.

PORC meetings were characterized = , by frank discussions and questioning of the proposed changes.

In particular, consideration was given to assure clarity and, consistency j among-procedures.

Items =for which adequate review time was not j - available-were postponed to allow committee members time for further j ' review and' comment.

Dissenting opinions were encouraged and resolved- ! to the satisfacti.on of the' committee prior to approval. The inspec - -torsiobserved that-PORC-adequately monitors and evaluates plant '

j i

performance-and conducts a thorough self-assessment-of; plant activi-1tiesfand programs.t ' = 7.2 iReview of Written Reports ' , , Periodic and special reports, andTitcensee event reports-(LERs) were.

,,

reviewed.for, clarity,' validity, accuracy of the root cause-and" safety.

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. significance description,- and adequacy of corrective actiont The % inspectors determined;whether further information'was required. The a

. inspectors'also verified that the reporting requirements of 10 CFR I o , .50.73; Station Administrat'ive and Operating Procedures;_and Technical A - , 4y Specification 6'.9 had beenimet. The-following' reports were reviewed: j - <q' . .

, LER.90-09 l Design Deficiency-Identified in CAR Fan Damper - Solenoid, Valves j , y LER 90-10 . Installed Test Equipment Results in Actuation of.

~ ^' ' Reactor Protection System x . ~ ." b LER 90-11 Potential-for loss of Sump Recirculation Due to Bus'- Undervoltage ' > 4m ' { \\ - l ,, NjCs ' _

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' jn, ' LER 90-12 Charging Pump Declared Inoperable Due to Gas Bindinq , LER 90-13 Incorrect 0-Ring Material Installed in Four Motor-Operated Valves , .LER 90-14 Incorrect Torque $ witch Settings on RCS Loop hotation ~ , Valves u 7' Haddam Neck Plant Monthly Operating Report 90-07, covering the period , July-I to Ju v 31, 1990, t' ' .New $witchgear Building Bimonthly Progress Report No. 23 - Final Report, dated August 10, 1990

Inoperable. Core Exit Thermocouple Accident Monitoring Instrumentation-5pecial-Report, dated August 10, 1990

Haddam Neck Plant Semiannual Fitness-for-Duty Performance Data, dated Auguat 17,L1990 i Inoperable-Emergency Diesel Generator Special Report, dated ( September 4 1990 ' i w No noteable observations were identified, ' , N 8.'. Exit Interviews; >

j" During this inspection, periodic. meetings-were held with station manage- ' 'is ment to'. discuss inspection observations and findings.. At the close of the '

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inspection period, an exit meeting was held to summarize the conclusions.= of: the-inspection. No written material was given to the: licensee'and no ,'g

proprietary information.related to

this inspection was identified.;

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